05000325/LER-2008-005, As-Found Values for Safety/Relief Valve Lift Setpoints Outside Technical Specification Allowed Tolerance
| ML082610226 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/10/2008 |
| From: | Wills E Progress Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 08-0119 LER 08-005-00 | |
| Download: ML082610226 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3252008005R00 - NRC Website | |
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Progress Energy SEP, 1:0 2008 SERIAL: BSEP 08-0119 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Brunswick Steam Electric Plant, Unit No. 1 Docket No. 50-325/License No. DPR-71 Licensee Event Report 1-2008-005 Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power
& Light Company, now doing business as Progress Energy Carolinas, Inc., submits the enclosed Licensee Event Report (LER). This report fulfills the requirement for a written report within sixty (60) days of a reportable occurrence.
Please refer any questions regarding this submittal to Mr. Philip A. Leich, Manager -
Support Services, at (910) 457-2271.
Sincerely, Edward L. Wills, Jr.
Plant General Manager Brunswick Steam Electric Plant MAT/mat
Enclosure:
- - Licensee Event Report Progress Energy Carolinas, Inc.
Brunswick Nuclear Plant P0 Box 10429 Southport, NC 28461
/7- ~-;9-.
Document Control Desk BSEP 08-0119 / Page 2 cc (with enclosure):
U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Luis A. Reyes, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Joseph D. Austin, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)
ATTN: Mrs. Farideh E. Saba (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2010 (9-2007)
, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE Brunswick Steam Electric Plant (BSEP), Unit 1 05000325 1 of 4
- 4. TITLE As-Found Values for Safety/Relief Valve Lift Setpoints Outside Technical Specification Allowed Tolerance
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR I SEQUENTIAL REV MONTH DAY YEAR FACILITY NAME DOCKET NUMBER MONH AYNEAUMBER NO.
MOT A
ER05000 FACILITY NAME DOCKET NUMBER 07 16 2008 2008 - 005 - 00 09 10 2008 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
LI 20.2203(a)(3)(i)
L-50.73(a)(2)(i)(C)
[] 50.73(a)(2)(vii) 1 El 20.2201(d)
Li 20.2203(a)(3)(ii)
[L 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
EL 20.2203(a)(1)
El 20.2203(a)(4)
[E 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
Li 20.2203(a)(2)(i)
El 50.36(c)(1)(i)(A)
E] 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
[j 50.36(c)(1)(ii)(A)
EL 50.73(a)(2)(iv)(A)
E] 50.73(a)(2)(x)
Li 20.2203(a)(2)(iii)
[L 50.36(c)(2)
El 50.73(a)(2)(v)(A)
[:1 73.71 (a)(4) 100 El 20.2203(a)(2)(iv)
[L 50.46(a)(3)(ii)
El 50.73(a)(2)(v)(B)
EL 73.71 (a)(5)
EL 20.2203(a)(2)(v)
EL 50.73(a)(2)(i)(A)
EL 50.73(a)(2)(v)(C)
Li OTHER El 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
[L 50.73(a)(2)(v)(D)
Specify in Abstract below or I
_in
Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
Introduction
On July 16, 2008, the Brunswick Steam Electric Plant as-found testing of 11 safety/relief valves (SRVs), which had been removed from Unit 1 during the spring 2008, refueling outage (i.e., BI 17R1) was completed. These results of the testing indicated that two of the 11 valves were found to lift at greater than 3 percent tolerance allowed by Technical Specification 3.4.3, "Safety/Relief Valves." A third SRV pilot valve was unable to be pressurized to obtain as-found data because of excessive disc-to-seat leakage.
Event Description
Initial Conditions At the time of the event, Unit 1 was in Mode 1, operating at approximately 100 percent of rated thermal power.
Discussion During the spring 2008, Unit 1 refueling outage, the 11 Model 7567F Target Rock Two-Stage pilot valve assemblies were replaced with certified spares. The removed SRV pilot valves were sent to Wyle Laboratories for set pressure testing. On July 16, 2008, as-found testing of the SRV pilot valves was completed. The testing indicated that two of the 11 valves actuated at pressures outside of the 3 percent tolerance allowed by Technical Specification 3.4.3. A third SRV pilot valve was unable to be pressurized to obtain as-found data because of excessive disc-to-seat leakage. The test data is provided in the following table.
Technical Valve Identification As-Found Specification Percent Difference (psig)
Setpoint (psig) 1-B21-FO13F 1177 1130 + 33.9
+4.16%
1-B21-FO13G Did Not Open 1130 + 33.9 N/A 1-B21-FO13J 1185 1150 +/- 34.5
+3.04%
Since Technical Specification 3.4.3 requires 10 of the 11 installed SRVs to be operable, this condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as operation prohibited by the plant's Technical Specifications.
Event Cause
The root cause of each SRV is discussed separately due to the individual nature of the findings.
SRV 1-B21-FO13F Inspections of the 1-B2 1-FO13F pilot valve assembly identified indications of pilot rod misalignment.
Additionally, the valve Was lapped for 120 seconds, which is longer than the typical limitation of 90 seconds. When the valve was lapped, a non-uniform or non-concentric seat band may have been formed.
This allowed the pilot disc to get cocked in the seat causing the misalignment in the pilot rod. Based on industry operating experience, pilot rod binding is known to cause setpoint drift. The root cause for this valve failure was an incomplete understanding of proper lapping techniques when the BSEP SRV rebuild program began and corrosion bonding resulting in setpoint drift.
SRV 1-B21-FO13G The excessive disc-to-seat leakage of this pilot valve caused steam cutting of the pilot disc and seat. The root cause for this valve failure was an incomplete understanding of proper lapping techniques when the BSEP SRV rebuild program began which allowed the pilot to leak.
SRV 1-B21-FO13J The leakage of this pilot valve caused steam cutting of the pilot disc and seat. The pilot seat was slightly ditched which causes a pinch between the pilot disc and seat. Corrosion bonding was also present. These factors combined to cause the setpoint drift. The root cause for this valve failure was an incomplete understanding of proper lapping techniques when the BSEP SRV rebuild program began which allowed the pilot to leak.
Safety Assessment
The safety significance of this condition is considered minimal. The as-found condition of the Unit 1 SRVs was compared to the current overpressure analysis prepared in support of extended power uprate and it was concluded that this analysis remained bounding. As such, the applicable acceptance criteria for design basis events would have been met and the SRVs remained capable of performing their intended safety function.
Corrective Actions
The 11 SRV pilot valve assemblies were replaced with certified spares during the BI 17R1 refueling outage.
To address lapping concerns, guidance from the Electric Power Research Institute (EPRI) Target Rock SRV Model 67F Maintenance Guide was incorporated into the SRV pilot rebuild procedure (i.e., OCM-VSR509).
This was completed as a corrective action for LER 1-2006-004, Supplement 1. Use of the lapping practices outlined in the guide will minimize ditching of the pilot disc to preclude failures as well as minimize pilot leakage.
The SRV pilot valves removed during the BI 17R1 refueling outage will be refurbished and recertified prior to reinstallation.
Previous Similar Events
A review of LERs and corrective action program condition reports for the past three years identified the following similar event.
LER 2-2007-003, dated August 16, 2007, as supplemented on October 18, 2007, and LER 1-2006-004, dated July 26, 2006, as supplemented on November 17, 2006, both document operation prohibited by TSs due to as-found testing which indicated that four SRVs actuated at pressures outside of the 3 percent tolerance allowed by Technical Specification 3.4.3. The failures observed following BI 17R1 are consistent with those experienced during the 2006 and 2007 refueling outages. These failures have been previously investigated (i.e., Nuclear Condition Reports (NCRs) 196311 and 237575. The root causes for those investigations remain valid and the previously identified corrective actions had not been completed for the set of pilot valves removed in 2008. As such, the current failures do not call into question the effectiveness of the prior
corrective actions
Commitments
No regulatory commitments are contained in this report.